A condensed review of nuclear reactor thermal-hydraulic computer codes for two-phase flow analysis

Abstract

A review is made of the computer codes developed in the U.S. for thermal-hydraulic analysis of nuclear reactors. The intention of this review is to compare these codes on the basis of their numerical method and physical models with particular attention to the two-phase flow and heat transfer characteristics. A chronology of the most documented codes such as COBRA and RELAP is given. The features of the recent codes as RETRAN, TRAC and THERMIT are also reviewed. The range of application as well as limitations of the various codes are discussed.